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1
Paramonova, Ekaterina (Ekaterina D. ). "CRUD resistant fuel cladding materials." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82447.
Full textAbstract:
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013."June 2013." Cataloged from PDF version of thesis.
Includes bibliographical references (pages 27-29).
CRUD is a term commonly used to describe deposited corrosion products that form on the surface of fuel cladding rods during the operation of Pressurized Water Reactors (PWR). CRUD has deleterious effects on reactor operation and currently, there is no effective way to mitigate its formation. The Electric Power Research Institute (EPRI) CRUD Resistant Fuel Cladding project has the objective to study the effect of different surface modifications of Zircaloy cladding on the formation of CRUD, and ultimately minimize its effects. This modification will alter the surface chemistry and therefore the CRUD formation rate. The objective of this study was to construct a pool boiling facility at atmospheric pressure and sub-cooled boiling conditions, and test a series of samples in simulated PWR water with a high concentration of nanoparticulate CRUD precursors. After testing, ZrC was the only material out of six that did not develop dark, circular spots, which are hypothesized to be the beginnings of CRUD boiling chimneys. Further testing will be needed to confirm that it is indeed more CRUD resistant, even under realistic PWR conditions in a parallel testing facility.
by Ekaterina Paramonova.
S.B.
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Reece, Warren Daniel. "Theory of cladding breach location and size determination using delayed neutron signals /." Diss., Georgia Institute of Technology, 1988. http://hdl.handle.net/1853/13317.
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Andrews, Michael Robert. "The interaction of deposition promoters with AGR fuel cladding surfaces." Thesis, University of Newcastle Upon Tyne, 1998. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.244466.
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Jarvis, Jennifer Anne. "Hydrogen entry in Zircaloy-4 fuel cladding : an electrochemical study." Thesis, Massachusetts Institute of Technology, 2015. http://hdl.handle.net/1721.1/103730.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 291-297).
Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.
by Jennifer Anne Jarvis.
Ph. D.
5
Lee, Youho. "Safety of light water reactor fuel with silicon carbide cladding." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/86866.
Full textAbstract:
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 303-314).
Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident conditions is examined with an integrated approach of experiments, modeling, and simulation. High temperature (1100°C~1500°C) steam oxidation experiments demonstrated that the oxidation of monolithic SiC is about three orders of magnitude slower than that of zirconium alloys, and with a weaker impact on mechanical strength. This, along with the presence of the environmental barrier coating around the load carrying intermediate layer of SiC fiber composite, diminishes the importance of oxidation for cladding failure mechanisms. Thermal shock experiments showed strength retention for both [alpha]-SiC and [beta]-SiC, as well as A1₂O₃ samples quenched from temperatures up to 1260°C in saturated water. The initial heat transfer upon the solid - fluid contact in the quenching transient is found to be a controlling factor in the potential for brittle fracture. This implies that SiC would not fail by thermal shock induced fracture during the reflood phase of a loss of coolant accident, which includes fuel-cladding quenching by emergency coolant at saturation conditions. A thermo-mechanical model for stress distribution and Weibull statistical fracture of laminated SiC cladding during normal and accident conditions is developed. It is coupled to fuel rod performance code FRAPCON-3.4 (modified here for SiC) and RELAP-5 (to determine coolant conditions). It is concluded that a PWR fuel rod with SiC cladding can extend the fuel residence time in the core, while keeping the internal pressure level within the safety assurance limit during steady-state and loss of coolant accidents. Peak burnup of 93 MWD/kgU (10% central void in fuel pellets) at 74 months of in-core residence time is found achievable with conventional PWR fuel rod design, but with an extended plenum length (70 cm). An easier to manufacture, 30% larger SiC cladding thickness requires an improved thermal conductivity of the composite layer to reduce thermal stress levels under steady-state operation to avoid failure at the same burnup. A larger Weibull modulus of the SiC cladding improves chances of avoiding brittle failure.
by Youho Lee.
Ph. D.
6
Paul, James. "Joining of silicon carbide for accident tolerant PWR fuel cladding." Thesis, University of Manchester, 2017. https://www.research.manchester.ac.uk/portal/en/theses/joining-of-silicon-carbide-for-accident-tolerant-pwr-fuel-cladding(f9851a0a-ef68-465e-8029-a31ab77fab27).html.
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Following two previous nuclear reactor accidents involving light water reactors, there is a renewed interest in accident tolerant fuels. These accident tolerant fuels should not oxidise in a steam atmosphere during loss of coolant accidents. One such accident tolerant fuel that has been suggested is the use of ceramic composite cladding material as a replacement for the current zircaloy cladding. The high temperature stability of silicon carbide, together with its high resistance to corrosion may make it preferable to zircaloy during accident conditions. Furthermore, if the neutron absorption cross section of the cladding is less than the current zircaloy, extended life might be available when compared with current fuels. One of the main difficulties in using ceramic cladding materials as nuclear fuels is the lack of a reliable joining process to manufacture end caps for the cladding tubes. A manufacturing method would need to be developed to produce ceramic joint that is able to withstand a PWR environment. Two methods of ceramic joining have been proposed. Firstly, silicon carbide deposition process that is used to infill the gap between two ceramic components and secondly a ceramic soldering technique. A silicon carbide deposition process has been developed. The deposit was confirmed to be 3C silicon carbide which has preferable irradiation response to the other polytypes. The deposit was found to be carbon rich which was largely removed through the use of a thermal treatment step. The deposit was used to coat metallic surfaces for increased hardness, reduced sliding wear and corrosion resistance. Silicon carbide joints were produced using an oxide powder frit of silicon dioxide, yttrium oxide and aluminium oxide. Tubular samples were joined, however they contained circumferential cracking resulting in a join that was not hermetically sealed. The thermal conductivity of each joint varied from sample to sample. X-ray computed tomography showed there were large inconsistencies in the volume of joined material present in each sample giving rise to the large variation in thermal conductivity.
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Phuah, Chin Heng. "Corrosion of thermally-aged Advanced Gas-Cooled Reactor fuel cladding." Thesis, Imperial College London, 2012. http://hdl.handle.net/10044/1/10550.
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The microstructure of Advanced Gas-Cooled Reactor (AGR) fuel cladding that underpins its corrosion behaviour has been established, contributing to an understanding of long-term control, monitoring practice and storage decisions for this fuel. AGR fuel cladding specimens sourced from Sellafield Ltd, cut and individually heat treated at temperatures from 400 to 800°C for 24 to 192h were attempts to approximate irradiated AGR fuel cladding and characterised both in terms of their corrosion behaviour and of microstructure. Niobium carbide (NbC) second phases are the primary local corrosion sites. Bulk austenite-γ cladding metal (50.3±1.7 at% Fe, 21.0±1.1 at% Cr and 21.0±0.4 at% Ni) around NbC precipitates exhibited extensive corrosion even though the precipitate themselves appear unchanged. Corrosion observed from the specimen surface took the form of lacy covers around an NbC precipitate at the cover centre (~10 to 25 μm dia. depending on the site) and in the subsurface were voids (~0.1 μm pin-holes), cavities (~2 to 5 μm), an envelope of dissolved-metal region along NbC peripheries (~1 μm thick with austenite-γ composition decreased by on average 20% Fe, 21% Cr and 17% Ni) or a large, smooth concave pit bottom comparable to the cover dimension. These observations collectively suggest that AGR cladding corrosion is a diffusion-controlled phenomenon where the NbC precipitate may act as the cathode in a local galvanic couple and the adjacent austenite-γ metal is the anode that undergoes preferential oxidation. The primary contributing factors to NbC-induced AGR cladding corrosion are high NaCl concentration of the electrolyte solution, large NbC precipitates, small austenite-γ grains and presence of stress in the microstructure. Specifically, corrosion potential measurements in the 0.001M electrolyte NaCl are ~800mV (v.s. Ag|AgCl reference electrode) more noble than in the 0.1M electrolyte, suggesting that cladding wet storage requires maintenance with lowest chloride concentration practically achievable. Specimens with comparatively large NbC precipitates (~5 μm) and small austenite-γ grains (~10 μm) that result from heat treatment are ~810mV more corrosion susceptible than the as-received specimens with ~0.1 dia. NbC precipitate and ~25 μm austenite-γ grains. Increased dislocation densities were observed adjacent to the grown-NbC precipitate, imparting a stress-corrosion effect on the AGR cladding corrosion.
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Rai, Subash. "Role of sulphur on carbon deposition on AGR fuel cladding steel." Thesis, University of Birmingham, 2019. http://etheses.bham.ac.uk//id/eprint/8882/.
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Filamentary carbon deposition on 20Cr/25Ni/Nb austenitic stainless steel from 1000-5000 ppm C2H4 in 1%CO/ bal. CO2 at 600°C has been investigated. This filamentary carbon deposition is catalysed by metallic nickel-rich particles formed intrinsically from the alloy during the early stages of oxidation. Samples were analysed using electron microscopy techniques. A simple model has been proposed to explain the formation of nickel-rich particles within the subsurface oxide layer. Furthermore, in this project the effect of COS, H2S and CH3SH on carbon deposition were examined. Addition of sulphur species suppressed the deposition process but the effect was dependent on the concentration of C2H4 and the sulphur species. This suppressing effect of sulphur on carbon deposition was explained in term of the role played by the adsorbed sulphur in reducing the rate of the key steps of the growth process of a filament.
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Carr, James. "Surface Modification Techniques for Increased Corrosion Tolerance of Zirconium Fuel Cladding." VCU Scholars Compass, 2016. http://scholarscompass.vcu.edu/etd/4474.
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Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively em- ployed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neu- tron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objec- tives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits.
This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and consid- ering materials and processes for modifying the surface of zircaloy fuel cladding. This innova- tive research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely.
Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their mi- crostructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffu- sion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spec- troscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide en- compassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.
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Jena, Anupam S. M. Massachusetts Institute of Technology. "Wettability of candidate Accident Tolerant Fuel (ATF) cladding materials in LWR conditions." Thesis, Massachusetts Institute of Technology, 2020. https://hdl.handle.net/1721.1/127299.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, May, 2020Cataloged from the official PDF of thesis.
Includes bibliographical references (pages [69]-70).
Since the 2011 Fukushima accident, substantial research has been dedicated to developing accident tolerant fuel (ATF) cladding materials. These analyses have mostly concentrated on the capability of potential ATF materials to withstand runaway steam oxidation and preserve their mechanical strength and structural integrity under thermal shocks. However, knowledge is relatively deficient for the thermal-hydraulic properties of these materials, particularly under light water reactor (LWR) operating conditions. The surface wettability is particularly important, as it affects the dynamics of the boiling heat transfer process, and consequently, the critical heat flux (CHF) and rewetting temperatures, which are important thermal limits for LWRs. Surface wettability determines nucleation site density, bubble departure diameter, and bubble departure frequency.
Therefore, it is essential to quantify the surface wettability of candidate ATF cladding materials to determine their thermal-hydraulic behavior compared to conventional Zircaloy claddings. The surface wettability is usually quantified through the sessile droplet contact angle, which is the angle formed between the liquid-vapor and the liquid-solid interface. The contact angle depends on the fluid, solid, surface finish, and operating conditions, i.e., temperature and pressure. However, most of the measurements available in the literature are performed at low pressure and in an inert atmosphere, which is quite different from the operating conditions of LWRs (i.e., in a steam-saturated atmosphere at a pressure as high as 15.5 MPa or 155 bars).
To close this gap, in this study, we designed and built an autoclave-type facility capable of measuring static, advancing, and receding contact angle in steam-saturated atmospheres, from sub-atmospheric conditions up to the critical point of water, i.e., 22.1 MPa (221 bar or 3200 psi) and 374°C. We measured the static contact angle of conventional Zircaloy-4 and candidate ATF cladding materials (e.g., Cr-coated Zr-4, FeCrAl, and SiC). The contact angle decreases with an increase in temperature for all the materials. Rough surfaces showed higher wettability, i.e., lower contact angle, compared to the smooth surfaces. These trends are expected from theory. All the materials showed different wettability under the same temperature and pressure conditions. Individual correlations for temperature dependence for each of them are proposed.
by Anupam Jena.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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ROSSI, JESUALDO L. "Efeito de tratamentos mecanotermicos na fluencia de aco inoxidavel austenitico estabilizado com niobio." reponame:Repositório Institucional do IPEN, 1987. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9872.
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Made available in DSpace on 2014-10-09T12:32:19Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:10:29Z (GMT). No. of bitstreams: 1 01540.pdf: 2534668 bytes, checksum: a9e18afc1ec717924aa15bdb9fad2b16 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Jones, John Rhys. "Prediction of PWR fuel cladding failure strain in a loss-of-coolant accident." Thesis, Imperial College London, 1986. http://hdl.handle.net/10044/1/38057.
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Stempien, John D. (John Dennis). "Behavior of triplex silicon carbide fuel cladding designs tested under simulated PWR conditions." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76948.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 101-107).
A silicon carbide (SiC) fuel cladding for LWRs may allow a number of advances, including: increased safety margins under transients and accident scenarios, such as loss of coolant accidents; improved resource utilization via a higher burnup beyond the present limit of 62 GWd/MTU; and improved waste management. The proposed design, referred to as Triplex, consists of three layers: an inner monolith, a central composite, and an outer environmental barrier coating (EBC). The inner monolith consists of dense SiC which provides strength and hermeticity to contain fission products. The composite layer is made of SiC fibers, woven around the monolith, and then infiltrated with a SiC matrix. The composite layer adds strength to the monolith and provides a pseudo-ductile failure mode. The EBC is a thin coating of SiC applied to the outside of the composite to protect it against corrosion. The ends of the tubes may be sealed via the bonding of SiC end caps to the SiC tube. Triplex tube samples, monolith-only samples, and SiC/SiC bonding samples (consisting of two blocks bonded together) were tested in three phases as part of an evaluation of the SiC cladding system. A number of samples were exposed to PWR coolant and neutronic conditions using an incore loop in the MIT research reactor (MITR-II). Other samples remained in their as-fabricated states for comparison. First, mechanical testing revealed significant strength reduction in the Triplex samples due to irradiation-induced point defects, corrosive pitting of the monolith, and possible differences in the behavior of the Triplex components. Some manufacturing abnormalities were also discovered which could have compromised strength. The Triplex samples tested here were not as strong as reported in a previous study. SEM analysis was able to follow the propagation of cracks from initiation, at the monolith inner surface, to termination, upon breaching the EBC. The composite layer was found to be key in dissipating the energy driving the crack formation. Second, three SiC/SiC bonding methods (six samples total) were tested in the MITR-II to 0.2 dpa, and five of the six samples failed. SEM analysis indicates radiation induced degradation of the bond material. Dimensional and volume measurements established the anisotropic swelling of the two SiC blocks in each bond sample, which would have caused shear stresses on the bonds, contributing to their failure. Finally, thermal diffusivity measurements of the Triplex samples show substantial decreases with irradiation (saturating at about 1 dpa) due to the accumulation of phonon-scattering defects and corrosion of SiC. By 1 dpa, the thermal diffusivity/conductivity of this SiC cladding design is diminished to a value lower than that of Zircaloy. In the as-fabricated state, a large difference exists between the monolith-only and Triplex samples due to the phonon scattering centers at the interfaces of the layers. With irradiation this difference decreases, suggesting that similar corrosion and radiation damage effects exist in both the monolith and Triplex samples.
by John D. Stempien.
S.M.
14
Al, Shater Abdulla Faisal. "Intergranular corrosion of sensitized 20Cr-25Ni-Nb stainless steel nuclear fuel cladding materials." Thesis, University of Manchester, 2010. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706485.
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Seshadri, Arunkumar. "Impact of reactor environment on quenching heat transfer of accident tolerant fuel cladding." Thesis, Massachusetts Institute of Technology, 2018. https://hdl.handle.net/1721.1/121711.
Full textAbstract:
This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018
Cataloged from student-submitted PDF version of thesis. Page 123 blank.
Includes bibliographical references (pages 106-116).
Development of accident tolerant fuels (ATF) for light water reactors (LWRs) came into focus for the nuclear engineering community after the accidents at Fukushima-Daiichi. The primary focus of the ATF program is to identify alternative fuel and cladding technologies that may provide enhanced safety, competitiveness, and economics. The new fuel design must also be compatible with present-day LWR design. For near-term applications, coatings on the nominal Zirconium-based cladding material and other metallic materials are being considered to improve the corrosion resistance and reduce the generation of hydrogen at high temperatures. Major ATF coating choices under consideration include chromium as a coating, iron-chromium-aluminum alloys (FeCrAl) as cladding and molybdenum as a coating, which have demonstrated better mechanical and oxidation behavior during the experimental testing.
Thermal-fluids characteristics are pivotal for a robust testing of ATF concepts as the proposed candidates may have an entirely different thermal-hydraulic behavior when compared to Zircaloy-4. ATF coatings may display very different boiling characteristics as a result of different microstructures and surface characteristics. In the present work, transient boiling heat transfer during quenching of the candidate ATF claddings on vertical rodlets is studied experimentally. The candidate ATF material (chromium, FeCrAl, and molybdenum) are applied on Zircaloy-4 rodlets. The vertical solid rodlets are heated to temperatures up to 1000 °C and are quenched in a saturated pool of water at atmospheric pressure. The temperature variation during the quenching of rodlets was recorded insitu with synchronized visualization of boiling regimes over the test specimen using a high-speed video camera.
The quench performance of the ATF coatings was analyzed based on the examination of various surface parameters such as wettability, roughness, emissivity and capillary wicking. In order to obtain a more realistic picture of the candidate performance during the emergency cooling reflood phase in a nuclear reactor, the coated rodlets are also oxidized in an autoclave before quenching. The performance of the candidate claddings is evaluated after oxidation and the surface characterized. It was observed from the post-test analysis that the surface characteristics and oxidation had a significant impact on the quench performance of ATF coatings, which varied between different coating materials. In order to better understand the thermal margins in a reactor specific environment, an analysis was performed on samples after exposing them to gamma rays. The gamma rays tend to change the surface wettability through a phenomenon called Radiation Induced Surface Activation.
A Gammacell 220E irradiator that uses 12 cobalt-60 pencil sources, arranged axially in a sample chamber at MIT, was used to irradiated the samples. The results of water quenching and contact angle studies showed a higher Leidenfrost temperature and wettability in both samples exposed to gamma irradiation. The detailed microscopic analysis attributed the enhanced wettability to oxidation of the surface under gamma irradiation.
by Arunkumar Seshadri.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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Konarski, Piotr. "Thermo-chemical-mechanical modeling of nuclear fuel behavior : Impact of oxygen transport in the fuel on Pellet Cladding Interaction." Thesis, Lyon, 2019. http://www.theses.fr/2019LYSEI080.
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L’objectif de cette thèse est d'étudier l'impact du transport de l’oxygène sur la thermochimie de l’interaction pastille-gaine. Pendant les rampes de puissance, le combustible nucléaire est exposé à des gradients de température élevés. Il subit des changements chimiques et structurels. Le gonflement du combustible entraîne un contact mécanique avec la gaine, provoquant des contraintes mécaniques élevées. Simultanément, des espèces chimiquement réactives sont libérées par le centre des pellets chauds et peuvent interagir avec la gaine. La combinaison de ces facteurs chimiques et mécaniques peut entraîner une défaillance de la gaine causée par la fissuration par corrosion sous contrainte. Il a été prouvé que le transport de l'oxygène sous des gradients de température élevés affecte la thermochimie, ce qui peut jouer un rôle important dans la fissuration par corrosion sous contrainte. Cette thèse présente des simulations 3D de rampes de puissance dans des réacteurs à eau sous pression avec le code de performance de combustible ALCYONE, qui fait partie de l'environnement informatique PLEIADES. Le code a été mis à jour pour associer la description de la thermochimie du carburant irradié déjà disponible au transport de l'oxygène en tenant compte de la thermo diffusion de l'oxygène. L'impact de la redistribution de l’oxygène pendant une période transitoire de puissance sur la thermochimie du combustible irradié et sur le relâchement de gaz chimiquement réactif provenant du combustible (I(g), I2(g), CsI(g), TeI2(g), Cs(g) et Cs2(g) principalement) est étudié. Les simulations montrent que la redistribution de l’oxygène, même modérée, conduit à la réduction des oxydes métalliques (dioxyde de molybdène, molybdates de césium, oxyde de chrome) au centre des pastilles de combustible et, par conséquent, au relâchement d’une quantité beaucoup plus importante de césium gazeux, en accord avec les examens post-irradiation. Les calculs tridimensionnels des quantités d'importance pour la fissuration par corrosion sous contrainte due à l'iode (contrainte circonférentielle, déformation circonférentielle, pression partielle d'iode sur gaine) sont ensuite utilisés dans des simulations de propagation de fissures de gaineThe goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation
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Auguste, Rasheed. "Quantifying the fouling resistance of Accident-Tolerant Fuel (ATF) cladding coatings with force spectroscopy." Thesis, Massachusetts Institute of Technology, 2017. http://hdl.handle.net/1721.1/112377.
Full textAbstract:
Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 418-420).
CRUD (Chalk River Unidentified Deposits) is buildup of metal oxides on the interior of nuclear reactors. This is caused by corrosion in reactor internals, leading to problems such as coolant contamination in porous deposits left by CRUD. CRUD has forced many nuclear reactors into temporary shutdown or production downgrades, costing millions of dollars US per reactor. If the CRUD growth factors could be fully understood, they could be controlled, and the CRUD problem could be eliminated altogether. Atomic force microscopy can be used to measure the force, or the strength of the CRUD-clad bond with different materials. This research focuses on answering this question: How does the force change between CRUD particles and different materials that could be used for reactor cladding? This study will analyze lab-grown CRUD samples on different substrate materials and characterize CRUD growth on each. It was found the CRUD-bond forces (from least to greatest) on silicon carbide (SiC), Titanium aluminum carbide (Ti2AlC), and max-phase zirconium alloy 211(Zr4M211) behaved similarly in air and in water. The forces on each surface increased with increasing dwell time for the Fe3O4 particle AFM tip; in contrast, most adhesion forces stayed constant with the NiO AFM tip. Furthermore, these CRUD forces were compared to other non-accident tolerant fuels, and there are cases in which non-ATF materials show more CRUD resistance (less adhesive force) than ATF-materials. This study's analysis could be applied to other materials to be used for reactor cladding. Once the material with the lowest-strength CRUD bond is identified and installed, the nuclear industry could save millions of dollars US per reactor fuel cycle.
by Rasheed Auguste.
S.B.
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Jernkvist, Lars Olof. "Modelling of pellet-cladding interaction induced failure of light water reactor nuclear fuel rods." Licentiate thesis, Luleå tekniska universitet, 1998. http://urn.kb.se/resolve?urn=urn:nbn:se:ltu:diva-26115.
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Li, Zhen. "SURFACE HARDENING OF AUSTENITIC FE–CR–NI ALLOYS FOR ACCIDENT-TOLERANT NUCLEAR FUEL CLADDING." Case Western Reserve University School of Graduate Studies / OhioLINK, 2018. http://rave.ohiolink.edu/etdc/view?acc_num=case150486174877088.
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McGrath, Margaret A. "The effect of an unstable niobium carbide population on the creep behaviour of AGR fuel cladding." Thesis, Open University, 1992. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.316615.
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Clark, Ronald N. "Mapping of corrosion sites in advanced gas-cooled reactor fuel cladding in long term pond storage." Thesis, Swansea University, 2018. https://cronfa.swan.ac.uk/Record/cronfa40783.
Full textAbstract:
This thesis aims to investigate sites where corrosion initiates on an austenitic stainless steel used to encapsulate advanced gas-cooled nuclear reactor fuel when in long term pond storage and subsequently provide insight into what happens when corrosion propagates. The combination of both a critical temperature and radiation dose within the reactor core may lead to very small quantities of this stainless steel becoming sensitised. The first chapter is a discussion on the literature and past study on the subject. Within the second chapter, analysis of the stainless steel in an as-received format is undertaken, and is relevant, given that only a very small quantity of cladding may be rendered sensitised. This chapter includes an investigation into Cl- concentration and temperature dependence. The third chapter focuses on identifying the initiation points of localised intergranular corrosion (IGC) in freely corroding and perturbed environments. An in-situ time-lapse microscopy and scanning vibrating electrode technique investigation are used in conjunction with post corrosion microscopy to study areas affected by IGC. The fourth chapter describes a higher resolution approach to investigating the initiation point of IGC. A scanning Kelvin probe force microscope is used to map the surfaces for areas in which corrosion may occur, while an atomic force microscope was used to provide high resolution images of tens of grains undergoing in-situ IGC. The aim of the final chapter is to provide an insight into reducing corrosion rates using an NaOH pre-treatment. NaOH is used in two different concentrations to study the effect on the free corrosion and breakdown potentials (Ecorr and Eb) on as-received and sensitised 20/25/Nb. This study will provide a scientic backbone that the industry may use to target areas of further research, helping provide ongoing safe storage of spent nuclear fuel in the UK.
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Gudipati, Mithun. "Computational fluid dynamics simulations of basket and fuel cladding temperatures within a rail cask during normal transport." abstract and full text PDF (free order & download UNR users only), 2007. http://0-gateway.proquest.com.innopac.library.unr.edu/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:1446432.
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Guenoun, Pierre S. M. Massachusetts Institute of Technology. "Design optimization of advanced PWR SiC/SiC fuel cladding for enhanced tolerance of loss of coolant conditions." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/103649.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 64-68).
Limited data has been published (especially on experimental work) on integrated multilayer SiC/SiC prototypical fuel cladding. In this work the mechanical performance of three unique architectures of three-layer silicon carbide (SiC) composite cladding is experimentally investigated under conditions associated with the loss of coolant accident (LOCA), and analytically under various conditions. Specifically, this work investigates SiC cladding mechanical performance after exposure to 1,400°C steam for 48 hours and after thermal shock induced by quenching from 1,200°C into either 100°C or 90°C water. Mechanical performance characteristics are thereafter correlated with sample architecture through void characterization and ceramography. The series with a reduced thickness did not have a pseudo-ductile regime due to overloading of the composite layer. The presence of the axial tow did not yield significant difference in the mechanical behavior most likely because samples were tested in the hoop direction. While as-received and quenched samples behaved similarly (pseudo ductile failure except for one series), non-frangible brittle failure (single-crack failure with no release of debris) was systematically observed after oxidation due to silica buildup in the inner voids of the ceramic matrix composite (CMC) layer. Overall, thermal shock had limited influence on sample mechanical characteristics and oxidation resulted in the formation of silica on the inner wall of the CMC voids leading to the weakening of the monolith matrix and brittle fracture. Stress field in the cladding design is simulated by finite element analysis under service and shutdown conditions at both the core's middle height and at the end of the fuel rod. Stresses in the fuel region are driven by the thermal gradient that creates stresses predominantly from irradiation induced swelling. At the endplug, constraints are mainly mechanical. Stress calculations show high sensitivity to the data scatter and especially swelling and thermal conductivity. No cladding with the design studied here can survive either service or shutdown conditions because of the high irradiation-induced tensile stresses that develop in the hot inner monolith layer. It is shown that this peak tensile stress can be alleviated by adjusting the swelling level of the different layers. The addition of an under-swelling material such as PyC or Si can reduce the monolith tensile stress by 10%. With a composite that swells 10% less than the monolith, the stress is reduced by 20%.
by Pierre Guenoun.
S.M.
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Bell, Benjamin. "The influence of alloying elements on the corrosion of Zr-based nuclear fuel cladding using density functional theory." Thesis, Imperial College London, 2016. http://hdl.handle.net/10044/1/51546.
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Zr-based alloys are used primarily as fuel cladding in water-cooled nuclear fission reactors. This is due to their good thermal and mechanical properties and low capture cross section for thermal neutrons. In this work, density functional theory (DFT) simulations were performed to investigate the behaviour of dopant elements in the cladding alloy oxide, with the aim of furthering the understanding of corrosion and hydrogen pick-up in Zr-based alloys. Simulations were performed in both monoclinic and tetragonal ZrO2 on single isolated defects and defect clusters, with the effect of compressive stress on some systems also simulated. Brouwer diagrams were constructed to model the interaction of multiple defect types within a system, in both equilibrium and non-equilibrium overall charge states. The concentration of the dopant elements in the oxide was fixed, and the chemical potential allowed to vary from the reference state, allowing the variation in oxidation state under varying oxygen partial pressure within the oxide to be investigated. Sn was shown to exist as Sn4+ under high oxygen partial pressures in tetragonal ZrO2, however at lower partial pressures a cluster with an oxygen vacancy (VO) and Sn2+ was the dominant defect type. As corrosion progresses, the oxygen partial pressure around a given immobile defect within the oxide layer will increase. Thus, as corrosion progresses the Sn2+:VO cluster will transition to Sn4+, with a consequent reduction in oxygen vacancy concentration. Oxygen vacancies help stabilise the tetragonal phase, and so this transition may cause a transformation from tetragonal to monoclinic to occur. The associated increase in volume (monoclinic phase has a volume ~4% larger) is proposed to be a contributing factor to the early transition observed in Sn-containing alloys. Nb was shown to exist in oxidation states ranging from Nb2+ to Nb5+ in the tetragonal phase, but only as Nb5+ in the monoclinic. In the lower oxidation states, Nb is able to mitigate the space charge which builds up in the oxide layer during corrosion, as a result of the lower oxygen vacancy diffusion rate compared to electrons. By mitigating the space charge, the corrosion kinetics are able to approach parabolic and the HPUF is lowered, both observations are in excellent agreement with experimental work which has shown Zr-Nb alloys to be unique in these properties. At transition, a large proportion of the tetragonal phase present in the oxide layer will transform to monoclinic as compressive stress is relieved. Since Nb was predicted to only exist in the 5+ state in the monoclinic phase, any lower states will oxidise upon phase transformation, releasing electrons into the oxide layer during transition. This oxidation process is proposed as a possible explanation for the sudden drop in HPUF that has been experimentally observed to occur around transition. Sc was shown to exist only as Sc3+ in both tetragonal and monoclinic phases, charge balanced by an increase in the concentration of oxygen vacancies. Since a stable cluster involving Sc3+ and VO would require the close proximity of two Sc3+ defects, it is assumed that at the low doping level in the model alloys produced alongside this work (0.2-0.4 %. wt.) stable clusters are unlikely to form in significant concentrations. Thus, the inclusion of Sc as a dopant increases the concentration of unbound (i.e. mobile) oxygen vacancies in the oxide layer. This is proposed as the reason for the extremely high corrosion rate observed in the Sc-containing model alloys produced and tested alongside this work. It is also assumed that the tetragonal phase fraction in the oxide layer of the Sc-containing alloys is likely to be high due to the increased oxygen vacancy concentration, however experimental testing to verify this prediction has not yet taken place. Sb was shown to exist as Sb5+ at high oxygen partial pressures, transitioning to Sb3+ at lower partial pressures in both the tetragonal and monoclinic phases. The application of a non-equilbrium charge state to the Brouwer diagrams showed that the oxygen partial pressure at which the transition between Sb5+ and Sb3+ occurs is able to smoothly change in order to counteract the applied space charge. This behaviour was observed in both phases, and implies that Sb may be able to act as a space charge compensation mechanism, in a similar fashion to Nb. The Zr-Nb-Sb model alloys produced alongside this work exhibited a lower corrosion rate and HPUF than the Zr-Nb alloys, suggesting that Sb may improve the beneficial effects already observed in Zr-Nb alloys.
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Fray, Elliott Shepard. "Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82456.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 176-179).
This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite steel which resists high temperature LBE corrosion. This composite steel consists of a Fel2Cr2Si protective layer weld overlaid on a T91 steel and then drawn to fuel cladding and piping material. A series of tests and materials analysis were performed on the composite piping material. These tests / analysis included microstructural characterization, heat treatment optimization, creep and tensile testing, diffusion testing, and long term static corrosion tests. Although the composite fuel cladding was not available at the time of this study, all of the results from the piping material characterization are directly applicable to the fuel cladding material. It has been shown that the heat treated composite piping material exhibits mechanical properties in excess of the ASTM minimum standard for T91. This material also exhibits a conservative corrosion rate of< 22pm/yr in static Pb-Bi eutectic. This low corrosion rate will enable fuel cladding to have a 3.6 year lifetime and piping material a 36 year lifetime, if the static corrosion rate is equivalent to the flowing corrosion rate. This material has also been shown to have a very slow diffusion rate for chromium, with a chromium inter-diffusion zone of < 35um over the lifetime of the nuclear reactor. There still however exist several challenges to implementing this technology. The challenges include resolving the issue of cracking of the Fel2Cr2Si layer during tube drawing and increasing the high temperature stress / creep resistance of the structural T91 layer.
by Elliott Shepard Fray.
S.M.
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Morgan, Andrew. "JOINING AND HERMETIC SEALING OF SILICON CARBIDE USING IRON, CHROMIUM, AND ALUMINUM ALLOYS." VCU Scholars Compass, 2014. http://scholarscompass.vcu.edu/etd/3529.
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Silicon Carbide (SiC) is increasingly gaining attention as a potential fuel cladding material, on account of its favorable thermo-mechanical and neutronic properties. The major limitations of such a cladding is currently associated with joining and hermetic sealing. The work presented here investigated the use of Al, Cr and Fe metals and a specialized alloy (FeCrAl) to achieve hermetic sealing of SiC tubes as well as a joining technology of SiC. Major part of solving this issue requires addressing joining of ceramic and metallic components, which are largely dissimilar in both thermal and mechanical properties. Preliminary experiments to bond SiC with FeCrAl resulted in adverse separation partially attributed to the differences in thermal expansion mismatch. To alleviate these problems, thin and thick coatings of the metals and alloys were applied to SiC. Qualitative microstructural characterization of the final product indicated satisfactory bonding between the materials.
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BORGES, JUNIOR REINALDO. "Desenvolvimento de método de medição das espessuras de núcleos e revestimentos de placas combustíveis." reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10607.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Matsson, Ingvar. "Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements." Doctoral thesis, Uppsala : Acta Universitatis Upsaliensis : Univ.-bibl. [distributör], 2006. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-6912.
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Haurais, Florian. "Evaluate the contribution of the fuel cladding oxidation process on the hydrogen production from the reflooding during a potential severe accident in a nuclear reactor." Thesis, Université Paris-Saclay (ComUE), 2016. http://www.theses.fr/2016SACLS375/document.
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En centrales nucléaires, un accident grave est une séquence très peu probable d’événements durant laquelle des composants du réacteur sont significativement endommagés, par interactions chimiques et/ou fusion, à cause de très hautes températures. Cela peut mener à des rejets radiotoxiques dans l’enceinte et à une entrée d’air dans le réacteur. Dans ce contexte, ce travail de thèse mené chez EDF R&D visait à modéliser la détérioration du gainage combustible, en alliages de zirconium, en conditions accidentelles : haute température et soit vapeur soit mélange air-vapeur. L’objectif final était d’améliorer la simulation par le code MAAP de l’oxydation du gainage et de la production d’hydrogène, en particulier pendant un renoyage avec de l’eau. Dû à l’épaississement progressif d’une couche de ZrO2 dense et protectrice, la cinétique d’oxydation du Zr en vapeur à hautes températures est généralement (sous-)parabolique. Cependant, à certaines températures, cette couche d’oxyde peut se fissurer, devenant poreuse et non protectrice. Par ce processus de « breakaway », la cinétique d’oxydation devient plus linéaire. De plus, l’augmentation de température peut mener les matériaux du réacteur à fondre et à se relocaliser dans le fond de cuve dont la rupture peut induire une entrée d’air dans le réacteur. Dans ce cas, l’oxygène et l’azote réagissent avec les gaines pré-oxydées, successivement par oxydation du Zr (épaississant la couche de ZrO2), nitruration du Zr (formant des particules de ZrN) et oxydation du ZrN (créant de l’oxyde et relâchant de l’azote). Ces réactions auto-entretenues relancent la fissuration du gainage et de sa couche de ZrO2, induisant une hausse de sa porosité ouverte. Afin de quantifier cette porosité du gainage, un protocole expérimental innovant en deux étapes a été défini et appliqué : il consistait à soumettre des échantillons de gainage en ZIRLO® à diverses conditions accidentelles pendant plusieurs durées puis à des mesures de la porosité ouverte par porosimétrie par intrusion de mercure. Les conditions de corrosion comprenaient plusieurs températures allant de 1100 à 1500 K ainsi que de la vapeur et un mélange air-vapeur 50-50 mol%. Pour les échantillons de ZIRLO® oxydés en vapeur, sauf à 1200 et 1250 K, les transitions de cinétique n’ont pas lieu et la porosité ouverte reste négligeable au cours de l’oxydation. Cependant, pour les autres échantillons, corrodés en air-vapeur ou oxydés en vapeur à 1200 ou 1250 K, des transitions « breakaway » sont observées et les résultats de porosimétrie montrent que la porosité ouverte augmente au cours de la corrosion, proportionnellement au gain en masse. De plus, il a été mis en évidence que la distribution de tailles de pores des échantillons de ZIRLO® s’étend significativement pendant la corrosion, en particulier après « breakaway ». En effet, ces tailles vont de 60 μm à environ : 2 μm avant la transition, 50 nm juste après et 2 nm plus longtemps après. Enfin, un modèle numérique en deux étapes a été développé dans le code MAAP pour améliorer sa simulation de l’oxydation du gainage. D’abord, grâce à la proportionnalité entre porosité ouverte et gain en masse des échantillons, des corrélations de porosité ont été implémentées pour chaque condition de corrosion. Ensuite, les valeurs de porosité calculées sont utilisées pour augmenter proportionnellement la vitesse d’oxydation du gainage. Ce modèle amélioré simule ainsi non seulement les réactions chimiques des gaines en Zr (oxydation et nitruration) mais aussi leur dégradation mécanique et son impact sur leur vitesse d’oxydation. Ceci a été validé en simulant des essais QUENCH (-06, -08, -10 et -16), conduits au KIT pour étudier le comportement de gaines dans des conditions accidentelles avec un renoyage final. Ces simulations montrent un meilleur comportement thermique du gainage et une production d’hydrogène significativement plus haute et donc plus proche des valeurs expérimentales, en particulier pendant le renoyageIn nuclear power plants, a severe accident is a very unlikely sequence of events during which components of the reactor core get significantly damaged, through chemical interactions and/or melting, because of very high temperatures. This may potentially lead to radiotoxic releases in the containment building and to air ingress in the reactor core. In that context, this thesis work led at EDF R&D aimed at modeling the deterioration of the nuclear fuel cladding, made of zirconium alloys, in accidental conditions: high temperature and either pure steam or air-steam mixture. The final objective was to improve the simulation by the MAAP code of the cladding oxidation and of the hydrogen production, in particular during a core reflooding with water. Due to the progressive thickening of a dense and protective ZrO2 layer, the oxidation kinetics of Zr in steam at high temperatures is generally (sub-)parabolic. However, at certain temperatures, this oxide layer may crack, becoming porous and not protective anymore. By this “breakaway” process, the oxidation kinetics becomes rather linear. Additionally, the temperature increase can lead core materials to melt and to relocate down to the vessel lower head whose failure may induce air ingress into the reactor core. In this event, oxygen and nitrogen both react with the pre-oxidized claddings, successively through oxidation of Zr (thickening the ZrO2 layer), nitriding of Zr (forming ZrN particles) and oxidation of ZrN (creating oxide and releasing nitrogen). These self-sustained reactions enhance the cracking of the cladding and of its ZrO2 layer, inducing a rise of its open porosity.In order to quantify this cladding porosity, an innovative two-step experimental protocol was defined and applied: it consisted in submitting ZIRLO® cladding samples first to various accidental conditions during several time periods and then to measurements of the open porosity through porosimetry by mercury intrusion. The tested corrosion conditions included numerous temperatures ranging from 1100 up to 1500 K as well as both pure steam and a 50-50 mol% air-steam mixture. For the ZIRLO® samples oxidized in pure steam, except at 1200 and 1250 K, the “breakaway” kinetic transitions do not occur and the open porosity remains negligible along the oxidation process. However, for all other samples, corroded in air-steam or oxidized in pure steam at 1200 or 1250 K, “breakaway” transitions are observed and the porosimetry results show that the open porosity increases along the corrosion process, proportionally to the mass gain. Moreover, it was evidenced that the pore size distribution of ZIRLO® samples significantly extends during corrosion, especially after “breakaway” transitions. Indeed, the detected pore sizes ranged from 60 μm down to around: 2 μm before the transition, 50 nm just after and 2 nm longer after. Finally, a two-step numerical model was developed in the MAAP code to improve its simulation of the cladding oxidation. First, thanks to the proportionality between open porosity and mass gain of cladding samples, porosity correlations were implemented for each tested corrosion condition. Second, the calculated porosity values are used to proportionally enhance the cladding oxidation rate. This improved model thus simulates not only chemical reactions of Zr-based claddings (oxidation and nitriding) but also their mechanical degradation and its impact on their oxidation rate. It was validated by simulating QUENCH tests (-06, -08, -10 and -16), conducted at KIT to study the behavior of claddings in accidental conditions with a final reflooding. These simulations show a better cladding thermal behavior and a hydrogen production significantly higher and so closer to experimental values, in particular during the reflooding
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Tioka, Jakub. "Výpočetní a experimentální analýzy jaderných paliv nové generace." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2021. http://www.nusl.cz/ntk/nusl-442550.
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The search for Accident tolerant fuels (ATF) which is the first part of this thesis is currently one of the most actual topics in the field of nuclear fuels. These fuels must be first successfully tested in operational and also accident conditions for their possible inclusion in commercial use. Following part of the thesis specifically focuses on the boiling crisis in nuclear reactors which can damage the nuclear fuel cladding. Therefore, it is necessary to know the critical heat flux value and the departure from nuclear boiling ratio. Calculations which determine critical heal flux value are placed in the practical part of the thesis. Calculations are compared with the data obtained during experiments. The ALTHAMC12 and the other correlations which are based on the previous measurements are used for the computational analysis.
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CASTANHEIRA, MYRTHES. "Analise dos mecanismos de degradacao de varetas combustiveis falhadas em reatores PWR." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11141.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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MUCSI, CRISTIANO S. "Proposição de um processo alternativo à fusão via forno VAR para a consolidação de cavacos prensados de zircaloy e estudo do sistema dinâmico do arco elétrico." reponame:Repositório Institucional do IPEN, 2005. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11400.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Halabuk, Dávid. "Zhodnocení termomechanického chování perspektivních jaderných paliv při havárii s vnosem reaktivity." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2016. http://www.nusl.cz/ntk/nusl-242902.
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The objective of this master’s thesis is to simulate thermo-mechanical behaviour of nuclear fuel in a pressurized water reactor during a reactivity initiated accident. An important part of this work is focused on examination of processes which occur during such accident and on creation of a detailed overview of material properties of nuclear fuel and fuel cladding which are necessary for simulations that closely reflect reality. Simulations in this thesis examine cases of fresh or irradiated nuclear fuel for two types of fuel cladding, Zircaloy-4, a material that is currently used in nuclear reactors, and ceramic matrix composite material made of SiC. The thesis also presents comparison of results with a corresponding international benchmark and an assessment of the influence of selected input parameters on obtained results.
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Pereira, Luiz Alberto Tavares. "Desenvolvimento de processos de reciclagem de cavacos de Zircaloy via refusão em forno elétrico a arco e metalurgia do pó." Universidade de São Paulo, 2014. http://www.teses.usp.br/teses/disponiveis/85/85134/tde-27052014-090225/.
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Reatores PWR empregam, como combustível nuclear, pastilhas de UO2 acondicionadas em tubos de ligas de zircônio, chamados de encamisamento. Na sua fabricação são gerados cavacos de usinagem que não podem ser descartados, pois a reciclagem deste material é estratégica quanto aos aspectos de tecnologia nuclear, econômicos e ambientais. As ligas nucleares têm altíssimo custo e não são produzidas no Brasil, sendo importadas para a fabricação do combustível nuclear. Neste trabalho são abordados dois métodos para reciclar os cavacos de Zircaloy. No primeiro, os cavacos foram fundidos utilizando um forno elétrico a arco para obter lingotes. O segundo usa a técnica da metalurgia do pó, onde os cavacos foram submetidos à hidretação e o pó resultante foi moído e isostaticamente prensado e, a seguir, sinterizado a vácuo. A composição química, as fases presentes e a dureza no material foram determinadas. Os lingotes foram tratados termicamente e laminados, sendo que as microestruturas foram caracterizadas por microscopia óptica e eletrônica de varredura. Os resultados para ambos os métodos mostraram que a composição do Zircaloy reciclado cumpre as especificações químicas e apresentaram microestrutura adequada para uso nuclear. Os bons resultados do método de metalurgia do pó sugerem a possibilidade de produzir pequenas peças, como as tampas do encamisamento - end-caps, usando a sinterização no formato quase final (near net shape).PWR reactors employ, as nuclear fuel, UO2 pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopies. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering.
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Venigalla, Venkata Vijaya Raghava. "Computational fluid dynamic simulations of natural convection/radiation heat transfer within the fuel regions of a truck cask for normal transport." abstract (free order & download UNR users only), 2007. http://0-gateway.proquest.com.innopac.library.unr.edu/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:1447694.
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Casella, Andrew M. "Modeling of molecular and particulate transport in dry spent nuclear fuel canisters." Diss., Columbia, Mo. : University of Missouri-Columbia, 2007. http://hdl.handle.net/10355/4695.
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Thesis (Ph. D.)--University of Missouri-Columbia, 2007.The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on November 26, 2007 Vita. Includes bibliographical references.
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Quaranta, Delphine. "Étude de voies potentielles pour le recyclage du zirconium des gaines en Zircaloy des combustibles nucléaires usés." Thesis, Toulouse 3, 2019. http://www.theses.fr/2019TOU30038.
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Le Zircaloy-4 est un alliage à base de zirconium (~ 98 % massique) constituant le gainage des assemblages nucléaires. Actuellement, les gaines de Zircaloy irradiées sont destinées au stockage géologique profond en raison de leur contamination en radioéléments (contamination issue du séjour en réacteur ainsi que du procédé de traitement). Elles sont classées en déchet de moyenne activité à vie longue suivant les recommandations de l'ANDRA (radioactivité : 10 6 - 10 9 Bq/g, périodes > 31 ans). Les gaines de Zircaloy irradiées représentent une part importante de l'inventaire des assemblages, ~ 25 % en masse. Le recyclage du zirconium contenu dans les gaines pourrait donc présenter un intérêt économique, soit dans le but de revaloriser le zirconium par refabrication de gaines (avec la contrainte imposée par la présence résiduelle de 93Zr), soit à minima pour déclasser le déchet coque en faible activité. Ce travail de thèse a pour objectif d'étudier les voies potentielles pour le recyclage du zirconium contenu dans les gaines de Zircaloy irradiées, et plus précisément l'électroraffinage en milieux de fluorures fondus. L'étude de la composition des gaines de Zircaloy des combustibles nucléaires usés a tout d'abord été réalisée afin d'identifier les radiocontaminants présents dans les gaines irradiées. Ces éléments sont soit des produits d'activation (Cr, Fe, Ni, Co, Sn, ...), soit des produits de fission (H, Sr (+ Y), Cs (+Ba), Eu, ...) et les actinides (U, Pu, Am et Cm). Une étude électrochimique des ions zirconium (IV) a ensuite été réalisée dans le milieu LiF-NaF. Elle a permis d'en déterminer les mécanismes de réduction en zirconium métallique. Puis, une étude classique de nucléation/croissance a également permis d'optimiser les conditions opératoires (i.e. nature de la cathode, concentration de ZrF4 dans le sel, densité de courant appliquée...) pour l'obtention d'un dépôt de zirconium métallique adhérant sur cathode solide inerte. La dernière partie de ce travail s'est focalisée sur l'électroraffinage de tronçons de Zircaloy "frais", soit avant passage dans le réacteur. Une attention particulière a été portée sur le comportement des constituants d'alliage (Fe, Cr et Sn) vis-à-vis de l'avancement de dissolution de l'anode. Bien que les électroraffinages se soient portés sur un matériau non radioactif, la combinaison des approches thermodynamique et expérimentale a permis de proposer un premier dimensionnement du cœur de procédé. L'ensemble de ce travail permet d'avancer un premier scénario pour le traitement des gaines de combustibles usésZircaloy-4 is an alloy mainly composed of zirconium (~ 98%wt.) constituting the cladding of nuclear assemblies. Currently, used Zircaloy claddings are intended for deep geological storage due to their contamination by radioelements from the nuclear reaction and the reprocessing process. They are classified as long-lived intermediate-level waste according to ANDRA recommendations (radioactivity: 10 6 - 10 9 Bq/g, periods > 31 years), as they represent 25%wt. of the assembly inventory. Zirconium recycling thus could present an economic interest, either to upgrade the zirconium by remanufacturing sheaths (with the constraint imposed by the residual presence of 93Zr), or to downgrade the cladding wastes into low activity waste. This thesis aims to study the potential routes for the recycling of zirconium contained in spent Zircaloy sheaths, and more precisely electrorefining in molten fluorides. The study of Zircaloy sheath composition of spent nuclear fuel was first carried out to identify the radioelements present in used claddings. These elements are either activation products (Cr, Fe, Ni, Co, Sn, etc.), or fission products (H, Sr (+ Y), Cs (+ Ba), Eu, etc.), or actinides (U, Pu, Am and Cm). An electrochemical study of the zirconium (IV) ions was carried out in LiF-NaF at 750 °C to determine its reduction mechanisms into metallic zirconium. Then, a nucleation / growth study was performed to optimize the operating conditions (ie nature of the cathode, concentration of ZrF4, current density applied, etc.), to obtain an adherent metal zirconium deposit on inert solid cathode. The last part of this work was focused on the electrorefining of "fresh" Zircaloy sections, i.e. before its stay in the reactor. Particular attention was paid to the behavior of the alloy constituents (Fe, Cr and Sn), during the electrolysis process. This work proposes a first scenario for the reprocessing of spent fuel claddings
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Wagih, Malik Mamoon AbdelHalim. "Fuel performance of multi-layered zirconium and silicon carbide based Accident Tolerant Fuel claddings." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119049.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 86-91).
The Accident Tolerant Fuel (ATF) program is focused on extending the time for fuel failure during postulated severe accidents compared to the standard UO₂-Zr alloy fuel system. This thesis investigates the feasibility of four different cladding concepts, two of which are zirconium-alloy based and two are SiC-based. The Zirconium-alloy based claddings are 1) Zr4-Chromium coated cladding and 2) Zr4-FeCrAl coated cladding with a molybdenum interlayer (Zr4-Mo/FeCrAl). The SiC-based claddings are 3) composite SiC coated with chromium (SiC/SiC-Cr) and 4) Three layered SiC cladding consisting of inner and outer monolith with a composite layer sandwiched in between (mSiC-SiC/SiC-mSiC). The coated claddings were kept to a 50[mu]m of coating thicknesses, deducted from the base layer thicknesses. The claddings were studied, using the multi-physics fuel performance tool MOOSE/BISON, under steady-state PWR operating conditions as well as two transients: power ramp and loss-of-coolant accident (LOCA). The major finding is that the chromium coated concepts proved to be the most promising in both Zr4 and SiC based claddings. The three layered SiC cladding showed a high probability of failure during normal operation and transient conditions, while the Zr4-Mo/FeCrAl cladding showed high plastic strains in the molybdenum layer making its possibilities of survival questionable. On the other hand, the Zr4-Cr and SiC/SiC-Cr concepts showed acceptable plastic strains for the chromium coatings, with the SiC/SiC-Cr being more advantageous during LOCA scenarios. Both concepts warrant further experimental investigation as well as modelling of beyond design-basis accidents.
by Malik Mamoon AbdelHalim Wagih.
S.M.
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Steckmeyer, Antonin. "Caractérisation et modélisation du comportement mécanique à haute température des aciers ferritiques renforcés par dispersion d'oxydes." Phd thesis, Ecole Nationale Supérieure des Mines de Paris, 2012. http://pastel.archives-ouvertes.fr/pastel-00819136.
Full textAbstract:
Le renforcement par dispersion d'oxydes nanométriques permet, d'une manière générale, d'améliorer la résistance mécanique des matériaux métalliques. Il autorise donc une augmentation de leur température maximale d'utilisation. De nombreux travaux de recherche sont menés au Commissariat à l'Énergie Atomique et aux Énergies Alternatives sur les aciers ODS, pour Oxide Dispersion Strengthened steels. S'inscrivant dans le cadre du développement du nucléaire civil de quatrième génération, ces travaux ont pour but de qualifier un matériau pouvant être utilisé en tant que matériau de gainage combustible à une température de 650℃. Ce travail de thèse a pour objectif d'améliorer la compréhension des propriétés mécaniques des aciers ODS, en cherchant d'une part à caractériser et à modéliser leur comportement en fluage, et d'autre part à caractériser leur nisotropie de comportement et à en identifier l'origine. Pour cela, de nombreux essais mécaniques ont été effectués entre 20℃ et 900℃ sur une nuance d'acier ODS ferritique de composition nominale Fe-14Cr1W0,26Ti + 0,3Y2O3 . Cette nuance a été élaborée au CEA, par mécanosynthèse puis extrusion à chaud, sous la forme d'une barre ronde. Les résultats obtenus démontrent la grande résistance mécanique à haute température de l'acier ODS étudié. Ils mettent également en évidence une forte dépendance de la ductilité et de la résistance du matériau vis-à-vis de la vitesse de sollicitation. Sur la base des différentes caractérisations expérimentales réalisées, un modèle de comportement macroscopique uniaxial a été développé. S'appuyant sur la description de trois écrouissages cinématiques et d'un terme de restauration statique, ce modèle démontre une capacité remarquable à reproduire le comportement mécanique du matériau en traction, en fatigue, en fluage et en relaxation. Par ailleurs, la caractérisation de l'anisotropie mécanique de la nuance d'acier ODS étudiée s'avère dépendre de la température. Deux modélisations polycristallines différentes ont été mises en place afin de reproduire cette anisotropie de comportement à partir des textures cristallographique et morphologique du matériau. Le désaccord observé entre les prévisions de ces modèles et les résultats expérimentaux conduit à formuler d'autres hypothèses sur la déformation des aciers ODS.
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Corpace, Fabien. "Soudage par résistance du gainage combustible ODS d'un réacteur nucléaire de 4ème génération." Phd thesis, Université Sciences et Technologies - Bordeaux I, 2011. http://tel.archives-ouvertes.fr/tel-00786263.
Full textAbstract:
Les alliages ODS (Oxide Dispersion Strengthened) sont des matériaux candidats pour la réalisation du gainage combustible des réacteurs nucléaires de Génération IV de type RNR-Na. Leurs propriétés mécaniques à haute température sont assurées par une dispersion d'oxydes nanométriques qui peut être modifiée sous l'effet de la température et notamment lors du passage par l'état liquide. Les procédés d'assemblage en phase solide sont donc préconisés. La méthode de soudage étudiée dans cette thèse est le soudage par résistance en bout. Une double approche simulation numérique-expérimentation a été mise en place pour étudier l'influence des paramètres opératoires sur le soudage et la soudure. L'ensemble des essais (expérimentaux et numériques) est réalisé selon la méthode des plans d'expériences factoriels complets. Un alliage ODS base fer à 20 % de chrome a été utilisé comme alliage d'essai.L'analyse de l'influence des paramètres opératoires sur le soudage est réalisée dans un premier temps. Les résultats montrent que l'étape de soudage peut se découper en trois phases. Dans une première phase, la température augmente au niveau des interfaces. Dans une seconde phase, le procédé de soudage est gouverné par le changement de section macroscopique lorsque le courant passe du bouchon massif à la gaine plus étroite. La température augmente alors dans la partie de gaine dépassant de l'électrode. Lorsque la température atteint une valeur suffisamment élevée, la partie de gaine dépassant de l'électrode se déforme, entrainant un affaissement général des pièces et constituant la troisième phase de l'étape de soudage. L'influence des paramètres opératoires sur les phénomènes physiques lors de l'opération d'assemblage est évaluée. Il est alors possible d'influer sur les sollicitations thermiques et mécaniques à l'aide des paramètres opératoires. L'analyse de l'influence des sollicitations thermomécaniques sur la soudure est réalisée dans un second temps. Sur les plages de paramètres opératoires étudiées, les résultats révèlent sur certaines soudures la présence de défauts de compacité ainsi qu'une modification de la microstructure et de la dispersion des oxydes. Les défauts de compacité peuvent être liés à des phénomènes thermiques et mécaniques au contact entre pièces. Les modifications de la microstructure sont liées à des phénomènes de recristallisation dynamique ou de fusion locale et donc à des phénomènes thermiques mais aussi mécaniques avec la présence de déformations importantes. Les modifications de la microstructure sont alors reliées à une modification de la dispersion d'oxydes. A l'aide des paramètres opératoires il est possible d'influer sur les températures et les déformations afin de limiter les modifications de la structure de l'alliage ODS ainsi que l'apparition de défaut de compacité. A l'aide de l'ensemble de ces résultats, la procédure de soudage est adaptée sur un alliage ODS à 9 % de chrome, nuance envisagée pour la réalisation des futurs gainages combustibles. L'influence des propriétés matériaux sur le soudage et la soudure est alors discutée en comparant les deux nuances ODS de compositions différentes mais aussi en comparant les résultats obtenus sur l'alliage ODS à 20 % de chrome avec un alliage non renforcé de composition semblable.
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Čásar, Ondřej. "Výpočet chování paliva reaktorů VVER programem FEMAXI-6." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2019. http://www.nusl.cz/ntk/nusl-400560.
Full textAbstract:
The submitted master thesis deals with finding the right combination of models describing the cladding behavior implemented in the FEMAXI-6 computational code and then comparing it with the benchmark Zaporoshye, Novovoronezh and the modified FERMAXI 6 program with implemented models describing the E110 behavior used for VVER-type reactor fuel rods. Initially, there is a description of the FEMAXI-6 nuclear fuel analysis program including its structure, calculation mechanics and input file description. Furthermore, the work presents the benchmarks used to evaluate individual combinations of fuel models. An important part is the description of the PIE measurement, including measurement uncertainty, which can affect the results of the work. The next chapter contains a description of what affects fuel behavior during irradiation, which can be mathematically expressed and modeled. The following chapter describes the material equations defining the E110 alloy used as cladding of the fuel rods and which were subsequently implemented into the FEMAXI-6 computing program. The last chapter is devoted to the description of the results with appropriate comments.
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Gurgen, Anil. "Estimation of coping time in pressurized water reactors for accident tolerant fuel claddings." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119048.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis. Page 105 blank.
Includes bibliographical references (pages 99-104).
The Fukushima Nuclear Power Plant (NPP) accident in Japan has motivated improving the safety of current light water reactors (LWRs). Accident tolerant fuels (ATF) are being developed to enhance the safety of LWRs by tolerating loss of active cooling in the core for a longer duration compared to standard UO₂ and Zirconium-based claddings. In this work, high-temperature steam oxidation characteristics of potential ATF claddings, monolayer iron-chromium-aluminum (FeCrAl) and Cr-coated Zircaloy, are experimentally investigated. Specifically, this work investigates the high-temperature oxidation characteristics of FeCrAl alloy after exposure to 1000-1400 °C steam for 1 hour. A model for oxidation of FeCrAl alloy was developed based on the measured weight gain. The severe degradation of the FeCrAl samples from the steam attack was observed at ~1400 °C. Experimental investigation of ATF claddings also included high-temperature oxidation of Cr-coated Zircaloy pressure tube. Post-test analysis showed that for some regions, the Cr-coating is still present after 90 minutes of exposure to 1200°C steam, protecting the Zircaloy substrate beneath the coated layer. The performances of the FeCrAl and Cr-coated ATF claddings under beyond design basis accidents (BDBA) are modeled with thermal-hydraulics design basis code TRACE. A 3-loop Pressurized Water Reactor (PWR) model is created and the following BDBAs are simulated for this study: large break loss of coolant accident (LOCA) without safety injection systems, short-term station blackout (SBO) without any mitigation actions from the beginning and long-term SBO with auxiliary feedwater flow for the first 24 hours and the no mitigation actions afterwards. Two models are used for high-temperature oxidation of FeCrAl: the MIT model based on the experimental results of this work, and the Oak Ridge National Laboratory (ORNL) model based on experimental results of ORNL's work. The results showed that ATF claddings increase the coping time and produce less hydrogen compared to Zircaloy cladding under the considered BDBAs scenarios.
by Anil Gurgen.
S.M.
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Le, Gloannec Brendan. "Modifications microstructurales sous sollicitations thermomécaniques sévères : application au soudage par résistance des gaines de combustibles en aciers ODS." Thesis, Bordeaux, 2016. http://www.theses.fr/2016BORD0367/document.
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Les aciers renforcés par dispersion d’oxydes (ODS) sont des matériaux candidats pourconstituer les gaines de combustibles des réacteurs nucléaires de IVème génération refroidis au sodium(RNR-Na). Le soudage par résistance en bout de ces gaines est étudié dans ce travail, avec pourobjectif la détermination et la compréhension des effets du procédé sur l’évolution de la microstructuredes aciers ODS à 9 % et 14 % de chrome à l’échelle des grains et des oxydes nanométriques. Uneapproche couplant caractérisations microstructurales sur soudures, simulation numérique et simulationphysique du procédé, à l’aide d’un simulateur thermomécanique Gleeble 3500, est adoptée. Le soudagepar résistance en bout impose localement des conditions de sollicitations sévères en termes dedéformation, vitesse de déformation et température. Ces sollicitations induisent un affinement de lamicrostructure dont l’origine peut être attribuée à un mécanisme de recristallisation dynamique (acier à14 % de chrome) ou à l’association de la recristallisation dynamique et des transformations de phase(acier à 9 % de chrome). Les conditions d’occurrence de la recristallisation dynamique sont étudiéessur ces matériaux. On montre notamment la possibilité d’une transition entre recristallisation dynamiquecontinue et discontinue sur l’acier à 14 % de chrome en fonction des conditions de sollicitation. Il estégalement observé que ces sollicitations thermomécaniques sévères induisent une augmentation de lataille des oxydes nanométriques associée à une diminution de leur fraction volumiqueOxide dispersion strengthened (ODS) steels are considered as candidate materials for thedevelopment of fuel cladding for sodium-cooled fast reactors (SFR). Resistance upset welding of thecladding is studied in this work. The aim is to determine and to understand the process effects on themicrostructure of ODS steels with 9% and 14% of chromium at the scales of the grains and thenanometric oxides. An approach coupling microstructural characterization of welds, numericalsimulation and physical simulation of the process, using a thermomechanical simulator Gleeble 3500,is proposed. Resistance welding locally imposes severe thermomechanical conditions in terms of strain,strain rate and temperature. Refinement of the microstructure is noted and correspond to a dynamicrecrystallization mechanism (14 % Cr steel) or the combination of dynamic recrystallization and phasetransformations (9 % Cr steel). The conditions of occurrence of dynamic recrystallization are studied.The possibility of a transition between continuous and discontinuous dynamic recrystallization is shownfor the 14 % Cr steel according to the loading conditions. Such severe thermomechanical conditionsinduce an increase in the size of nanoscale oxides associated with a decrease of their volume fraction
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Drieux, Patxi. "Elaboration de tubes épais de SiC par CVD pour applications thermostructurales." Phd thesis, Université Sciences et Technologies - Bordeaux I, 2013. http://tel.archives-ouvertes.fr/tel-00958465.
Full textAbstract:
L'objectif de la thèse était de synthétiser des tubes de SiC monolithiques pour améliorer l'étanchéité de la structure composite SiC/SiC d'une gaine de combustible nucléaire. Des revêtements tubulaires de 8 mm de diamètre et quelques centaines de micromètres d'épaisseur ont été produits par dépôt chimique en phase vapeur à pression atmosphérique à partir d'un mélange CH3SiHCl2/H2. Le procédé a été développé de manière à réaliser en continu des tubes de SiC de plusieurs dizaines de centimètres de long. La composition chimique et la microstructure des tubes ont été déterminées par microsonde de Castaing, spectroscopie Raman, DRX et microscopie électronique (MEB, MET). Les propriétés mécaniques des tubes ont été caractérisées par nanoindentation et à travers des essais de compression C-ring. Le comportement thermomécanique a également été étudié. L'étude du procédé comprend une étude thermocinétique, un suivi de la phase gazeuse par IRTF et la modélisation 2D du réacteur.
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PEREIRA, LUIZ A. T. "Desenvolvimento de processos de reciclagem de cavacos de Zircaloy via refusão em forno elétrico a arco e metalurgia do pó." reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23302.
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Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-01-21T10:18:45Z
No. of bitstreams: 0Made available in DSpace on 2015-01-21T10:18:45Z (GMT). No. of bitstreams: 0
Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Sukjai, Yanin. "Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/87496.
Full textAbstract:
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 285-288).
There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 °C. The search for an accident tolerant cladding intensified after the Fukushima events of 2011. Silicon carbide (SiC) possesses several desirable characteristics as fuel cladding in light water reactors (LWRs). Compared to zirconium, SiC has higher melting point, higher strength at elevated temperature, and better dimensional stability when exposed to radiation, as well as lower thermal expansion, creep rate, and neutron absorption cross-section. However, under irradiation, the thermal conductivity of SiC is degraded considerably. Furthermore, lack of creep down towards the fuel causes the fuel-cladding gap and gap thermal resistance to stay relatively large during in-core service. This leads to higher fuel temperature during irradiation. In order to reduce the high fuel temperature during operation, the following fuel design options were investigated in this study: using beryllium oxide (BeO) additive to enhance fuel thermal conductivity, changing the gap bond material from helium to lead-bismuth eutectic (LBE) and adding a central void in the fuel pellet. In addition, the consequences of using thorium oxide (ThO₂) as host matrix for plutonium oxide (PuO₂) were covered. The effects of cladding thickness on fuel performance were also analyzed. A steady-state fuel performance modeling code, FRAPCON 3.4, was used as a primary tool in this study. Since the official version of the code does not include the options mentioned above, modifications of the source code were necessary. All of these options have been modeled and integrated into a single version of the code called FRAPCON 3.4-MIT. Moreover, material properties including thermal conductivity, swelling rate, and helium production/release rate of BeO have been updated. Material properties of ThO₂ have been added to study performance of ThO₂-PuO₂ . This modified code was used to study the thermo-mechanical behavior of the most limiting fuel rod with SiC cladding, and explore the possibility to improve the fuel performance with various design options. The fuel rod designs and operating conditions of a 4-loop Westinghouse pressurized water reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were chosen as representatives of conventional PWRs and upcoming SMRs, respectively. Sensitivity analyses on initial helium gap pressure, linear heat generation rate (LHGR) history, and peak rod assumptions have been performed. The results suggest that, because of its lower thermal conductivity, SiC is more sensitive to changes in these parameters than zirconium alloys. For a low-conducting material like SiC, an increase in cladding thickness plays a significant role in fuel performance. With a thicker cladding (from 0.57 to 0.89 mm), the temperature drop across the cladding increases, which makes the fuel temperature higher than that with the thin cladding. Reduction of fuel volume to accommodate the thicker cladding also causes negative impact on fuel performance. However, if the extra volume of the cladding replaces some coolant, the reduced coolant fraction design (RCF) has superior performance to the decreased fuel volume fraction design. In general, the most effective fuel temperature improvement option appears to be the option of mixing beryllium oxide into the fuel. This method outperforms others because it improves the overall thermal conductivity and reduces the overall temperature of the fuel. With lower fuel temperature, fission gas release and eventually plenum pressure -- one of the most life-limiting factor for SiC -- can be lowered.
by Yanin Sukjai.
S.M.
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Schnier, Gladys. "Development and validation of a fatigue-resistant cladding technology." Thesis, University of Strathclyde, 2015. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=26011.
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Knowledge and understanding of residual stresses has been and remains a complex area of study, the determination of these stresses crucial in appreciating the state of the component both prior to and during operation. In the case of dissimilar joints, the interaction of materials increases the complexity of the residual stress state resulting due to the applied joining process. A fatigue-resistant cladding technology is presented in this thesis, with the aim of inducing compressive residual stresses in the clad layer. Through the generation of these beneficial compressive residual stresses an improvement in fatigue performance can be achieved in an erosive-corrosive environment. Characterisation of clad and substrate materials allows an understanding of the interaction of materials and the resulting residual stress distribution due to a weld cladding process, both experimentally and through finite element modelling. Finite element modelling of the weld cladding process utilised an elastic-perfectly plastic material model throughout the investigation of the modelling process. Good correlation between experimental and simulation residual stresses is presented, with factors influencing residual stress distributions discussed. The accurate capturing of residual stresses due to weld cladding is a complex process due to, for example, material metallurgy and properties. Laser cladding provides an alternative method of generating residual stresses indicating that the fatigue-resistant cladding technology is not limited to weld cladding. Furthermore, weld cladding is not a process that must be applied exclusively, with the application of autofrettage post-cladding providing a means of favourably modifying tensile residual stresses obtained through the weld cladding process. The fatigue-resistant weld cladding technology has been successfully developed and validated, with recommendations provided for further development and the implementation of this concept. Although this research was primarily focussed on the weld cladding of a hydraulic fracturing pump, the application of this technology is not limited to this component therefore presenting the potential to improve the fatigue performance of any component operating under cyclic loading conditions in an erosive-corrosive environment.
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Parga, Clemente José. "Correlation between microstructures and oxidation resistance in Zr-Nb-Ti alloys." To access this resource online via ProQuest Dissertations and Theses @ UTEP, 2009. http://0-proquest.umi.com.lib.utep.edu/login?COPT=REJTPTU0YmImSU5UPTAmVkVSPTI=&clientId=2515.
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Baurens, Bertrand. "Couplages thermo-chimie mécaniques dans le dioxyde d'uranium : application à l' intéraction pastille-gaine." Thesis, Aix-Marseille, 2014. http://www.theses.fr/2014AIXM4047/document.
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En rampe de puissance, le combustible nucléaire est soumis à d'importantes contraintes thermiques et mécaniques, et subit une modification profonde de son environnement chimique. Le combustible contraint fortement la gaine, notamment au niveau des zones inter-pastilles, ce qui, associé au relâchement de produits de fission corrosifs, peut conduire à sa rupture par corrosion sous contraintes. Les évolutions simultanées de la mécanique, de la thermique et de la chimie du combustible sont liées, et participent au bon ou mauvais comportement de l'UO2 en rampe de puissance. L'objectif de ce travail est de modéliser à l'échelle d'une pastille de combustible, l'évolution couplée de la chimie, de la thermique et de la mécanique, et de préciser l'impact de ces couplages sur le comportement de l'UO2 en rampe de puissance. La finalité est d'évaluer un terme source en relâchement d'iode pour alimenter les modèles de corrosion sous contraintes dédiés aux études d'Interaction Pastille-GaineNuclear fuels under power transient undergo high thermal and mechanical stresses, as well as deep chemical modifications. Stresses on the cladding at the inter-pellet plane due to the pellet thermal expansion, associated to the corrosive fission product release, can lead to clad failures, resulting from a stress corrosion cracking mechanism. The thermal, mechanical and chemical properties of the UO2 irradiated fuel are closely dependent and play a major role on the behavior of the material during a power transient. The aim of this work is to model at the pellet scale the chemical, thermal and mechanical coupled changes of the UO2 fuel during a power transient scenario and to evaluate the consequences on the fuel behavior. The final objective is to obtain an evaluation of the iodine release source term to be used in I-SCC modelling codes dedicated to Pellet-Clad-Interaction studies
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Mabrouki, Mohamed. "Caractérisation de la tenue mécanique des assemblages bouchon-gaine en acier ODS obtenus par soudage par résistance." Electronic Thesis or Diss., Bordeaux, 2024. http://www.theses.fr/2024BORD0044.
Full textAbstract:
Les aciers ODS (Oxide Dispersion Strengthened) ferrito-martensitiques, font partie des matériaux candidats à la fabrication des pièces pour le gainage de combustible des réacteurs nucléaires de 4ième génération. L’assemblage type « bouchon-gaine » se fait par le procédé de soudage par résistance (SPR) en bout ; procédé de soudage en phase solide, connu pour avoir un impact limité sur la dispersion des nano-oxydes dans la zone soudée par rapport aux procédés de soudage par fusion. Un des objectifs de ces travaux est d’évaluer et de comprendre les effets du SPR sur la résistance mécanique finale de l’assemblage « bouchon-gaine » en acier 11Cr-ODS. Une approche couplant caractérisations microstructurales et mécaniques avec des simulations numériques (du procédé de SPR et des essais mécaniques) est adoptée. L’originalité de cette approche réside aussi dans le développement de deux géométries spécifiques d’éprouvettes de traction, permettant de localiser les sollicitations dans la zone soudée. En effet, les sollicitations thermomécaniques sévères induites par le SPR génèrent des hétérogénéités microstructurales au matériau avec des conséquences directes sur la résistance mécanique. Des microstructures complexes en termes de taille de grain, de texture locale, de phases en présence (ferrite, martensite, ferrite résiduelle) et de type de grain (recristallisé ou déformé) sont obtenues. Les essais mécaniques réalisés ont mis en évidence que la résistance mécanique de l’assemblage soudé peut être principalement associée à la zone interne du plan de joint, formant un angle d’environ 45° par rapport à l’axe de la gaine. Cette zone subit une déformation plastique significative, présente les valeurs de dureté les plus élevées et affiche un affinement plus prononcé de la microstructure. Un second objectif est l’évaluation des effets d’un traitement thermique post-soudage sur les propriétés microstructurales de la soudure et sur la résistance mécanique de l’assemblage soudé. Son effet est significatif s’il est réalisé au-dessus de la température de transformation de phase Ac3, tandis qu’il est limité, s’il est effectué en dessous d’Ac3. Lors d’essais de traction à température ambiante, la zone de rupture est déplacée de la zone soudée vers le métal à l’état de réception lorsque l’assemblage a subi un traitement thermique post-soudage adéquatOxide Dispersion-Strengthened (ODS) ferrito-martensitic alloys are among the candidate materials for the manufacture of fuel cladding parts for 4th generation nuclear reactors. The « plug-clad » assembly is carried out by the Pressure Resistance Welding (PRW) process; a solid phase welding process known to have a limited impact on the dispersion of nano-oxides in the welded zone compared with fusion welding processes. One of the aims of this work is to assess and understand the effects of PRW on the final mechanical strength of the 11Cr-ODS steel plug-clad assembly. An approach coupling microstructural and mechanical characterizations with numerical simulations (PRW process and mechanical tests) is adopted. The originality of this approach also lies in the development of two specific geometries for tensile samples, enabling the localization of stresses in the welded zone. Indeed, the severe thermomechanical loadings imposed on the material during the PWR process generate microstructural heterogeneities in the material with direct consequences on its mechanical resistance. Complex microstructures in terms of grain size, local texture, phases (ferrite, martensite, residual ferrite) and grain type (recrystallized or deformed) are obtained. The mechanical tests indicate that the mechanical resistance of the welded assembly is primarily associated with the internal zone of the joint plane, forming an angle of approximately 45° with respect to the axis of the clad. This area is submitted to significant plastic deformation, presents the highest hardness values, and exhibits a more pronounced refinement of the microstructure. A second objective is the evaluation of the effects of a post-welding heat treatment on the microstructural properties of the weld and on the mechanical strength of the welded assembly. Its effect is significant if it is carried out above the phase transformation temperature, Ac3, while it is limited if below Ac3. During tensile tests at room temperature, the fracture zone is moved from the welded area to the as-received metal when the assembly has undergone adequate heat treatments